Author: Mehmet MercimekPublisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Development of a Nodal Method for the Solution of the Neutron Diffusion Equation in Cylindrical Geometry
Author: Mehmet MercimekDevelopment of a Nodal Method for the Solution of the Neutron Diffusion Equation in General Cylindrical Geometry
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
The usual strategy for solving the neutron diffusion equation in two or three dimensions by nodal methods is to reduce the multidimensional partial differential equation to a set of ordinary differential equations (ODEs) in the separate spatial coordinates. This reduction is accomplished by "transverse integration" of the equation.1 For example, in three-dimensional Cartesian coordinates, the three-dimensional equation is first integrated over x and y to obtain an ODE in z, then over x and z to obtain an ODE in y, and finally over y and z to obtain an ODE in x. Then the ODEs are solved to obtain onedimensional solutions for the neutron fluxes averaged over the other two dimensions. These solutions are found in regions ("nodes") small enough for the material properties and cross sections in them to be adequately represented by average values. Because the solution in each node is an exact analytical solution, the nodes can be much larger than the mesh elements used in finite-difference solutions. Then the solutions in the different nodes are coupled by applying interface conditions, ultimately fixing the solutions to the external boundary conditions.
A Nodal Integral Method for the Neutron Diffusion Equation in Cylindrical Geometry
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
This Summary reports recent progress in deriving and numerically implementing a nodal integral method (NIM) for solving the neutron diffusion equation in cylindrical, r-z, geometry. Comparisons of numerical solutions to two test problems with those obtained by the code EXTERMINATOR-2 indicate the superior accuracy of the nodal integral method solutions on much coarser meshes. 6 refs., 1 fig., 1 tab.
Energy Research Abstracts
Author: Nuclear Science Abstracts
Author: Handbook of Nuclear Engineering
Author: Dan Gabriel CacuciPublisher: Springer Science & Business Media
ISBN: 0387981306
Category : Science
Languages : en
Pages : 3701
Book Description
This is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all levels, this book provides a condensed reference on nuclear engineering since 1958.
Nuclear Science Abstracts
Author: Numerical Methods and Techniques Used in the Two-dimensional Neutron-diffusion Program PDQ-5
Author: L. A. HagemanPublisher:
ISBN:
Category : FORTRAN (Computer program language)
Languages : en
Pages : 90
Book Description
Government Reports Annual Index
Author: Publisher:
ISBN:
Category : Government reports announcements & index
Languages : en
Pages : 1396
Book Description
Nodal Method for Three-dimensional Fast-reactor Calculations in Hexagonal Geometry. [LMFBR].
![Nodal Method for Three-dimensional Fast-reactor Calculations in Hexagonal Geometry. [LMFBR]. PDF](https://books.google.com/books/content/images/frontcover/wJRGAQAACAAJ?fife=w80-h100)
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
A nodal method is developed for the solution of the multigroup neutron-diffusion equation in three-dimensional hexagonal-z geometry. The method employs an extension to hexagonal geometry of the transverse-integration procedure used extensively in the development of nodal schemes in Cartesian geometry. The partially-integrated fluxes in the three hex-plane directions are approximated by a polynomial tailored to the unique properties of the transverse-integrated equations in hexagonal geometry. The final equations, which are cast in the form of local inhomogeneous response matrix equations for each energy group, involve spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node.